• 한국압력기기공학회 논문집
• /
• 제10권1호
• /
• pp.44-50
• /
• 2014

There were many flaw issues of reactor vessel head penetration in USA fleets. USNRC issued 10CFR50.55a to implement reactor vessel head penetration ultrasonic examination performance demonstration(PD) in US for enhancement of inspection reliability. After September 2009, all US utilities inspected their RVHP with PD qualified system. Korea Hydro and Nuclear Power Company(KHNP) have developed reactor vessel head penetration performance demonstration system for ultrasonic test to apply for pressurized light-water reactor power plants in accordance with 10CFR50.55a since September 2011. RVHP configuration surveying and analysis, code requirement analysis, and performance demonstration specimen design were performed up to this day. Fingerprinting of manufactured specimen, development of test data management program, development of operation procedure, input of flawed data, and development of final report will be performed for the next step. This paper describes the development status of the performance demonstration system for reactor vessel head penetration ultrasonic examination in Korea.

• 한국압력기기공학회 논문집
• /
• 제6권2호
• /
• pp.9-15
• /
• 2010

US NRC issued rulemaking of 10CFR50.55a to perform the Perservice and Inservice inspection for Reactor Vessel Head Penetration Nozzle of US Nuclaer plant. The rulemaking was required the EPRI Demonstration to verify the NDE technique performing special Ultrasonic examination. In order to meet this requirement, the UT and ECT procedures was demonstrated and the NDE personnel were qualified by EPRI. In this paper, the NDE technique and analysis method are described the Preservice inspection for the Palo Verde #1/2/3 Replacement Reactor Vessel Head Penetration Nozzle using the qualified procedures and personnel.

• 한국전산구조공학회논문집
• /
• 제27권3호
• /
• pp.191-198
• /
• 2014

본 논문에서는 노심용융사고 시 관통노즐이 제거된 원자로용기 하부헤드의 구조 건전성 평가를 수행하였다. 열응력, 노심용융물의 질량 그리고 내압조건의 해석결과를 고려할 때, 하부헤드의 열응력에 의한 영향이 가장 크게 나타났다. 손상 가능성은 파손기준에 따라 평가하였으며, 등가소성변형률이 임계변형률 파손기준보다 낮은 수준으로 평가되었다. 열-구조물 연성해석 결과 하부헤드의 두께 중간층에서 항복강도보다 낮은 응력이 발생한 탄성영역 구간을 확인하였다. 내압이 커지면서 탄성영역 범위가 점차 좁아지면서 탄성영역이 내벽으로 이동하는 결과를 확인하였고, 노심용융사고 시 구조적 건전성을 만족하는 것으로 평가되었다.

• Nuclear Engineering and Technology
• /
• 제49권7호
• /
• pp.1524-1536
• /
• 2017

During the fabrication process of reactor vessel head penetration (RVHP), the grain size of the tube material can be changed by hot or cold work and the inner side of the tube can also be shrunk due to welding outside of the tube. Several nonregular time-of-flight diffraction (TOFD) signals were found because of deformed grains. In this paper, an investigation of nonregular TOFD indications acquired from RVHP tubes using experiments and computer simulation was performed in order to identify and distinguish TOFD signals by coarse grains from those by Primary Water Stress Corrosion Crack (PWSCC). For proper understanding of the nonregular TOFD indications, microstructural analysis of the RVHP tubes and prediction of signals scattered from the grains using Finite Element Method (FEM) simulation were performed. Prediction of ultrasonic signals from the various sizes of side drilled holes to find equivalent flaws, determination of the size of the nonregular TOFD indications from the coarse grains, and experimental investigation of TOFD signals from coarse grain and shrinkage geometry to identify PWSCC signals were performed. From the computer simulation and experimental investigation results, it was possible to obtain the nonregular TOFD indications from the coarse grains in the alloy 690 penetration tube of RVHP; these nonregular indications may be classified as PWSCC. By comparing the computer simulation and experimental results, we were able to confirm a clear difference between the coarse grain signal and the PWSCC signal.

• 대한방사선방어학회:학술대회논문집
• /
• 대한방사선방어학회 2009년도 춘계 학술발표
• /
• pp.304-305
• /
• 2009

• 대한용접접합학회:학술대회논문집
• /
• 대한용접접합학회 2005년도 춘계학술발표대회 개요집
• /
• pp.185-187
• /
• 2005

The reactor pressure vessel head of PWR has penetrations for control rod drive mechanism and instrumentation systems. The Primary coolant water and operating temperature can cause the stress-corrosion cracking of these nickel-based alloy penetrations. It is difficult to detect and size flaws such as SCC in the reactor head penetrations using conventional W methods because of complex geometry, Therefore, the utilities are using the TOFD technique for the detection and sizing of the flaw. This study shows the correlation between the ultrasonic wave direction and the orientation of the flaw and the range of flaw depth which can be detected by the TOFD techniques.

• 대한기계학회논문집A
• /
• 제37권9호
• /
• pp.1159-1168
• /
• 2013

가압형 경수로 원자로의 압력용기 상부헤드 관통노즐 J-groove 용접부 주변에서 일차수응력부식균열(PWSCC)로 인한 냉각수 누설사례가 발생하고 있다. 본 연구에서는 PWSCC 의 주요 원인 중 하나인 용접 잔류응력을 유한요소 해석을 이용해 평가하고 원자력 발전소의 정상가동 조건을 해석에 반영하는 방법이 용접잔류응력 분포에 미치는 영향에 대한 분석을 수행하였다. 또한 반복되는 원자력 발전소의 가동 주기가 용접잔류응력 분포에 미치는 영향을 확인하여 정상가동조건에서의 정확한 용접 잔류응력을 예측할 수 있는 방법을 분석하였다.

• 한국압력기기공학회 논문집
• /
• 제9권1호
• /
• pp.56-61
• /
• 2013

This paper presents the results of integrity assessment for the cracks happened in reactor vessel upper head penetration nozzles. The crack morphology for a boat sample from crack area was analyzed through microscope. The stress condition including weld residual stress around crack was analyzed using finite element analysis. From the results of crack morphology and stress condition, the crack was concluded as primary water stress corrosion cracking. The integrity of the cracked nozzle was assessed by the methodology provided in ASME Section XI. According to the assessment results, the remaining life of the cracked nozzle was 1.43 yrs. and the plant decided to repair it.

• 한국공작기계학회:학술대회논문집
• /
• 한국공작기계학회 2005년도 춘계학술대회 논문집
• /
• pp.216-227
• /
• 2005

The head penetrations for control rod drive mechanism and instrumentation systems are installed at the reactor pressure vessel head of PWRs. Primary coolant water and the operating conditions of PWR plants can cause cracking of these nickel-based alloy through a process called primary water stress corrosion cracking (PWSCC). Inspection of the head penetrations to ensure the integrity of the head penetrations has been interested since reactor coolant leakages were found at U. S. reactors in 2000 and 2001. The complex geometry of the head penetrations and the very low echo amplitude from the fine, multiple flaws due to the nature of the see made it difficult to detect and size the flaws using conventional pulse-echo UT methods. Time-of-flight-diffraction technique, which utilizes the time difference between the flaw tips while pulse-echo does the flaw response amplitude from the flaw, has been selected for this inspection for it's best performance of the detection and sizing of the head penetration see flaws. This study defines the limits of the detectable and accurately sizable minimum flaw size which can be detected by the General TOFD and the Delta TOFD techniques for circumferentially and axially oriented flaws respectively. These results assures the reliability of the inspection techniques to detect and accurately size for various kind of flaws, and will also be utilized for the future development and qualifications of the TOFD techniques to enhance the detecting sensitivity and sizing accuracy of the flaws of the reactor head penetrations in nuclear power plants.

• 비파괴검사학회지
• /
• 제36권4호
• /
• pp.273-280
• /
• 2016

비파괴검사 분야에 대한 시뮬레이션은 다양한 결함에 대한 신호의 예측과 검사 절차 개발에 사용되어진다. 특히 비파괴검사 전용 시뮬레이션 툴인 CIVA는 정확도가 높고 빠른 계산이 가능하며, 비파괴평가 기술과 동일한 형태의 화면 표시와 시각적으로 개선된 3차원 그래픽 유저 인터페이스를 제공한다. CIVA 소프트웨어 개발자가 내부적으로 타당성 검증을 시행하겠지만, 사용 이전에 소프트웨어의 정확도를 평가하는 독립적인 유효성 검증 연구가 필요하다. 이러한 목적으로 이번 연구에서는 CIVA를 이용하여 원자로 상부 헤드 관통관 검사에 사용되는 보정시험편에 대하여 TOFD 신호를 시뮬레이션하고, 실제 검사 신호와 비교하여 시뮬레이션 신호의 정확도와 적용 범위에 대하여 검증하였다. 종합적으로, A-scan 신호, B-scan 이미지, 깊이 측정 측면에서 CIVA 시뮬레이션 결과와 실험 결과 간에 전반적으로 일치를 보였다.